Korean Nuclear Power Plants Research Articles

Logistical Simulation Modeling for Planning a Soil Remediation Process

This paper describes the development of a discrete event simulation model using the FlexSim software to support planning for soil remediation at Korean nuclear power plants that are undergoing decommissioning. Soil remediation may be required if site characterization shows that there has been radioactive contamination of soil from plant operations or the decommissioning process. The simulation model was developed using a dry soil separation and soil washing process. Preliminary soil data from the Kori 1 nuclear power plant was used in the model. It was shown that a batch process such as soil washing can be effectively modeled as a discrete event process. Efficient allocation of resources and efficient waste management including volume and classification reduction can be achieved by use of the model for planning the soil remediation process. Cost will be an important criterion in the choice of suitable technologies for soil remediation but is not included in this conceptual model.

Remaining and emerging issues pertaining to the human reliability analysis of domestic nuclear power plants

Abstract Probabilistic safety assessments (PSA) have been used for several decades to visualize the risk level of commercial nuclear power plants (NPPs). Since the role of a human reliability analysis (HRA) is to provide human error probabilities for safety critical tasks to support PSA, PSA quality is strongly affected by HRA quality. Therefore, it is important to understand the underlying limitations or problems of HRA techniques. For this reason, this study conducted a survey among 14 subject matter experts who represent the HRA community of domestic Korean NPPs. As a result, five significant HRA issues were identified: (1) providing a technical basis for the K-HRA (Korean HRA) method, and developing dedicated HRA methods applicable to (2) diverse external events to support Level 1 PSA, (3) digital environments, (4) mobile equipment, and (5) severe accident management guideline tasks to support Level 2 PSA. In addition, an HRA method to support multi-unit PSA was emphasized because it plays an important role in the evaluation of site risk, which is one of the hottest current issues. It is believed that creating such a catalog of prioritized issues will be a good indication of research direction to improve HRA and therefore PSA quality.

Characteristics of the vertical and horizontal response spectra of the Gyeongju earthquake (12 September 2016, ML = 5.8) series

ABSTRACTThe response spectra of the 20 Gyeongju earthquake series, including the main earthquake (12 September 2016; ML = 5.8), were compared with the design response spectra (Regulatory Guide 1.60, 1973; United States Nuclear Regulatory Commission) for Korean nuclear power plants and the Korean Building Code (KBC) of 2016 for general structures and buildings. A total of 720 horizontal and vertical accelerations were used to derive the response spectra after normalisation by reference to peak ground acceleration. The horizontal and vertical response spectra were strongly dependent on the epicentral distance and the resonance frequency. When developing design response spectra, it is necessary to consider distances >200 km from the epicenter, given the strong dependence of the response on distance. The vertical responses were consistently lower than the horizontal responses, in good agreement with the various existing seismic design standards and empirical rules. It is essential to develop site-specific seismic design standards, reflecting local seismotectonic environments. Such efforts require large amounts of data covering earthquakes of various magnitudes in the Korean Peninsula and surrounding areas, and such standards should replace Regulatory Guide 1.60. In terms of the KBC 2016, it is necessary to make the criteria more conservative in the short periods, and then to create comprehensive design response spectra by referencing large amounts of qualitative data from the region around the Korean Peninsula on earthquakes of various magnitudes and their effects at different distances from the epicenter.

Comparison of Wall Thinning Analysis Results between CHECWORKS and ToSPACE

Assumptions have always been that wall thinning on the secondary side piping in nuclear power plants is mostly caused by Flow-Accelerated Corrosion (FAC). Recent studies have showed that wall thinning on the secondary side piping is caused by Liquid Droplet Impingement Erosion (LDIE), Solid Particle Erosion (SPE), cavitation, and flashing. To manage those aging mechanisms, several software such as CHECWORKS, COMSY, and BRT-CICERO have been used in nuclear power plants. Korean nuclear power plants have been using the CHECWORKS program since 1996 to date. However, many site engineers have experienced a lot of inconveniences and problems in using the CHECWORKS program. In order to work through the inconveniences and to remedy problems, KEPCO-E&C has developed a “3D-based pipe wall thinning management program (ToSPACE)” based on the experience of over 30 years in relation to the pipe wall thinning management. This study compares the results of FAC and LDIE analysis using both the CHECWORKS and ToSPACE programs with respect to validation of the wall thinning analysis results.

Analysis of CRUD Flake Applied to Abnormal High Beam Current by Shielded-EPMA

CRUD specimens, scraped from twice-burned fuel cladding in the Korean Nuclear Power Plant, were analyzed using Shielded-EPMA. The principal elements of the CRUD were identified as Ni and Fe, at an approximate ratio of 1.3 Ni/Fe. To investigate the morphology and composition of the pure metallic materials in the CRUD, coolant impurities must be removed. This can be accomplished by increasing the EPMA current to an abnormally high intensity until the impurities are melted. Normally, EPMA applications are performed at conditions of 20 kV voltage and 20 nA current. But in our study, the applied current was increased up to 1200 nA, over time increments ranging from 5 to 30 seconds. This technique was performed by opening an adjustable aperture for the gun alignment. Results showed impurities contained in the CRUD material disappeared and pure metal materials, e.g., Ni and Fe, remained. This method presents an innovative way to analyze CRUD.

Development of an Adaptive Underexcitation Limiter in Excitation System

A generator is designed to operate in the underexcitation region within its capability according to the system conditions. However, a sustained underexcitation condition owing to the partial and complete loss-of-field in the generator can cause thermal damage and system instability. Thus, an automatic voltage regulator and an underexcitation limiter should control the underexcitation condition of the generator, and the loss-of-field relay should be utilized to guarantee reliable operation. Therefore, in this paper, we discuss the coordination of generator underexcitation protection among various factors. In addition, in order to enhance the conventional underexcitation limiter, we proposed an adaptive underexcitation limiter. Through this method, a possibility of miscoordination can be avoided and the operation is modified adaptively according to the loss-of-field condition. The performance of the proposed method was verified via a 345-kV Korean nuclear power plant and transmission system using the Electromagnetic Transients Program.

Simulation of a laboratory-scale experiment for wave propagation and interaction with a structure of undersea topography near a nuclear power plant using a divergence-free SPH

The Divergence-Free SPH (DFSPH) method is a recently proposed novel incompressible Smoothed Particle Hydrodynamics (SPH) method. The DFSPH method enforces incompressibility via two iterative solvers: the Divergence-Free (DF) solver and the Constant-Density (CD) solver. In this study, the DFSPH algorithm is implemented into the SOPHIA Plus framework to simulate a set of wave propagation under the same geometry and conditions of a laboratory-scale experiment. The experiments are conducted for wave propagation and interaction with a structure of scaled-down undersea topography near the Kori nuclear power plant in South Korea. This study compares the free surface propagation and the wave height with the experimental measurements for three test cases: low/medium/high frequency waves. Overall, the simulation shows good agreement with the experiment both qualitatively and quantitatively. However, according to the sensitivity study, more realistic water splashing behaviors are captured as the particle size is reduced. The predicted wave heights at three different locations are also in fairly good agreement with the experimental measurement. Slight differences are observed after the wave collides with the structure because of the low energy dissipation in the simulation. However, the differences are not significant.

The effects of personality types on self-reported safety behavior: Focused on plant workers in Korea

We sought to validate a safety behavior tool used in South Korean nuclear power plants, and to investigate the effects of HEXACO personality types on safety behaviors. The participants were 242 individuals employed in corporate safety management who answered the questionnaires on safety behaviors checklist, impulsiveness, affectivity, job burnout, and perfectionism. An exploratory factor analysis was conducted on the safety behavior items, and the convergent and discriminant validity were confirmed through correlational analyses with the existing related variables. To examine the individual effects of personality variables on the validated safety behavior questionnaire, we introduced control variables into a subsequent hierarchical regression analysis. The analyses revealed that the personality variables had significant effects on the subscales of the safety behavior scale. The present study is significant in that it revealed that personality, a broad construct, can predict human errors and safety behaviors, which have had previously been found to associate with only specific variables, such as stress, impulsiveness, and perfectionism.

Development of a quantitative resilience model for nuclear power plants

The causes of the Fukushima nuclear power plant (NPP) accident have been identified as not only technical factors such as the structure, system, and equipment design, but also inadequate management of the human and organizational factors, which were the major contributors to exacerbating the beyond design basis accident. After the accident, the safety paradigm was changed to address the failure of equipment as well as effective factors for safety. Resilience engineering is a relatively new paradigm for safety management that focuses on how to cope with complexity under pressure or disturbance to achieve success, addressing the limitations of existing safety analysis measures. This study aimed to develop a quantitative resilience model for the NPP using a modified NPP resilience model based on the Model of Resilience in Situation developed by the Électricité de France. Event reports for Korea NPPs were analyzed according to the perspectives of Safety-I and Safety-II, and statistical analysis was performed to identify the relations in the resilience model. Through this analysis, the quantitative relationship of the element in the resilience model was determined, and the quantitative resilience model was developed. The developed quantitative resilience model was also validated through a statistical method. Our results provide a new method for safety assessment in NPPs, which can complement the conventional safety assessment. The proposed method is expected to be an index for evaluating the integrity of safety management in Korean NPPs.

비상디젤발전기의 한계상태 내진성능 평가

Generally in South Korea, nuclear power plants have been designed in accordance with the Regulatory Guide 1.60 design spectra which effective frequency range is between 2 and 10 Hz in. However, for CEUS in hard rock sites, effective frequency could be higher than 10 Hz. The CEUS and Korean nuclear power plant sites were showed similar. The possibility of failure for sensitive power plant components in the high frequency range should be considered and evaluated in those sites. In this study, a shaking table test with emergency diesel generator (EDG) was conducted in order to improve the reliability of nuclear plant and administrative control procedure. From the test results, the leak of fuel and fuel inlet form bolt connection and fuel filter were observed. The check list of nuclear plant component after an earthquake can be included in bolt connections of EDG as well as anchor bolt.

Increased risk of thyroid cancer in female residents nearby nuclear power plants in Korea: was it due to detection bias?

BackgroundThe Korea Radiation Effect & Epidemiology Cohort - The resident cohort (KREEC-R) study concluded that there is no epidemiological or causal evidence supporting any increase in cancer risks resulting from radiation from Korean nuclear power plants (NPPs). But the risks of thyroid cancer in women were significantly higher in residents living near NPPs than control. Debate about the cause of the pattern of thyroid cancer incidence in women is ongoing and some researchers argue that detection bias influenced the result of KREEC-R study. Therefore there was a need to investigate whether residents living near NPPs who were assessed in the KREEC-R were actually tested more often for thyroid cancer. We evaluated the possibility of detection bias in the finding of the KREEC-R study based on materials available at this time.MethodsUsing the KREEC-R raw data, we calculated age standardized rates (ASRs) of female thyroid cancer and re-analyzed the results of survey on the use of medical services. We also marked the administrative districts of residents who received the Radiation Health Research Institute (RHRI) health examinations and those in which thyroid cancer case occurred as per the Chonnam National University Research Institute of Medical Sciences (RIMS) final report on maps where the locations of NPPs and 5 km-radii around them were also indicated. And we compared the incidence rates of Radiation-induced cancer measured between the first period when RHRI health examinations were not yet implemented, and the second period when the RHRI health examinations were implemented.ResultsThe ASR for the far-distance group, which comprised residents living in areas outside the 30 km radius of the NPPs, increased rapidly after 2000; however, that of the exposed group, which comprised residents living within a 5 km radius of the NPPs, started to increase rapidly even before 1995. The frequencies of the use of medical services were significantly higher in the intermediate proximate group, which comprised residents living within a 5–30 km radius of the NPPs, than in the exposed group in women. In case of female thyroid cancer, the second period ASR was higher than the first period ASR, but in case of female liver cancer and female stomach cancer no significant difference were observed between the periods. On map, many administrative districts of residents who received RHRI health examinations and most administrative districts in which thyroid cancer case occurred on RIMS final report were outside 5 km-radii around NPPs.ConclusionsWe could not find any evidence supporting the assertion that detection bias influenced the increased risks of female thyroid cancer observed in the exposed group of the KREEC-R study, as opposed to the control group.

Experimental study to assess effects of ballooning and fuel relocation on the coolability of fuel rod bundle

For a large break loss-of-coolant accident (LB LOCA) in a pressurized water reactor, the clad temperature increases until the reflood phase. This causes ballooned fuel rods and fuel relocation, thereby reducing the flow passage area of the subchannels and redistributing the flow and heat transfer in the subchannels. The fuel relocation induces augmentation of power and temperature at the ballooned region simultaneously. In this light, research on the coolability of the deformed fuel and fuel relocation is very important for preventing and mitigating severe accidents. In the present study, reflood experiments were performed to investigate the coolability in the ballooned fuel rod geometry and fuel relocation condition under a LB LOCA for a Korean nuclear power plant, APR1400. The coolability experiments were carried out using three kinds of test sections. The first experiments using a 6 × 6 intact rod bundle were performed as a reference case. The second experiments simulated the deformed fuel using a 5 × 5 ballooned rod bundle. The last experiments were performed using another ballooned 5 × 5 rod bundle with a local power increase to simulate fuel relocation. The reflood phenomena were scrutinized for these three rod bundles. While there were not substantial differences in the peak cladding temperature between the intact and ballooned cases, the fuel relocated bundle showed high peak cladding temperatures.

원전지역공동체 재구조화에 대한 시론적 연구: 고리원전지역공동체 확장과정에 대한 개념적 접근

본 연구는 고리원전 지역사회가 지난 반세기 가까운 시기에 겪었던 여러 가지 변화와 관련된 외부적･정책적 사건흐름을 기술하고자 한다. 특히 고리지역원전공동체가 고리원자로 주변지역(부산광역시 기장군 장안읍 길천마을)으로부터 부산 전역까지 어떻게 지리적･사회적 외연을 확장하였으며, 그 확장과정에 영향을 미친 맥락적(외부적･정책적) 사건은 무엇이 있었는가를 밝히는 작업이 본 연구의 주된 연구내용이다. 이러한 연구 작업을 통하여 다양한 원전하위공동체의 존재와 그 특성을 확인할 수 있었다. 이러한 연구결과들이 각 원전지역공동체와 그 공동체 구성원들에 상당한 영향(위험, 갈등, 편익 등)을 미치는 원전관련 의사결정과정과 문제해결과정에 이들이 의미 있는 결정주체로 참여하여 각각의 의견과 요구를 반영할 있는 단초를 제공하기를 기대한다. 본 논문은 원전지역공동체와 관련하는 5가지 경험요소들(위험경험, 편익경험, 원전관련 지식, 원전관련활동, 물리적 거리)의 수준이나 경험 강도에 따른 3가지 원전지역공동체의 개념적 재구성을 시도하였다. 기존에 고리원전인접지역으로 물리적으로 한정된 지역공동체를 장안원전지역공동체, 기장원전지역공동체, 부산원전지역공동체등으로 개념적 확장을 하였으며, 각각의 공동체의 특성을 확인하는 작업을 진행하였다. 아울러 원전지역공동체 확장과정에 개입했다고 판단되는 주요 맥락적 사건들(고리원전1호기 정전사태, 후쿠시마사고, 기장해수담수화사업, 경주지진 등)과 공동체 확장과정과의 관계를 살펴보았다. 연구결과는 새롭게 등장하고 그 영향력을 강하게 발휘하기 시작한 다른 원전지역공동체들의 성격이나 그 공동체들이 제기하는 요구, 구성원들의 정서를 이해하고 이에 능동적으로 대응하고 포용하는 노력이 필요하다는 점을 시사하고 있다. 이러한 장기간에 걸친 흐름을 파악하기 위한 연구방법으로 본 연구는 관련 문헌연구와 관련 전문가들에 대한 초점집단면접방법(Focused Group Interview)을 활용하였다.

Modeling Safety-II based on unexpected reactor trips

Safety-I is defined as a state where as few things as possible go wrong. Until now, the methods for analyzing the safety in nuclear power plants (NPPs), i.e., Probabilistic Safety Assessment and Deterministic Safety Analysis, have been developed from the perspective of Safety-I. However, focusing solely on Safety-I may miss opportunities to 1) learn from successes, and 2) observe how human adaptation contributes to successful outcomes, despite novel circumstances and resource limitations. In this light, a paradigm shift from ensuring that “as few things as possible go wrong (Safety-I)” to ensuring that “as many things as possible go right (Safety-II)” has been suggested. This study aimed to develop a model of Safety-II for unexpected situations in NPPs. Safety-II concentrates on the conditions in which as many things as possible go right. First, this study suggested and characterized a qualitative Safety-II model, which was modified from a resilience model developed by the Électricité de France (EDF). Second, event reports from unplanned reactor trips at Korean NPPs were analyzed based on the elements of this characterized Safety-II model as well as event severity. Third, a quantitative network model of Safety-II was developed by performing a correlation analysis. Finally, a feasibility analysis of Safety-I and Safety-II concepts for explaining event severity was performed. The result of this research suggests a new methodology for the safety assessment of unexpected reactor trips in NPPs, which could complement the conventional probabilistic safety assessments and deterministic safety analyses.

Methodology for determination of alarm and warning set-points for radioactive effluent monitors in Korean pressurized water reactors

All radioactive gaseous and liquid effluents discharged from Korean nuclear power plants are monitored by effluent monitors to prevent effluent releases to the environment under uncontrolled conditions. This paper provides the methodology and parameters used in the calculation of alarm (high) and warning (low) set-points for gaseous and liquid effluent monitors in Korean pressurized water reactors. Alarm set-points are determined to assure compliance with the Korean regulatory limits of concentration of radioactive effluents. Even though warning set-points are not required by the regulatory body, Korean pressurized water reactors determine the warning set-points of effluent monitors not only to take an active management of effluent discharge but also to keep radiation doses to members of the public living around nuclear power plants as low as reasonably achievable.

Necessity of management for minor earthquake to improve public acceptance of nuclear energy in South Korea

Abstract As public acceptance of nuclear energy in Korea worsens due to the Fukushima accident and the earthquakes that occurred in the Gyeongju area near the Wolsong nuclear power plant (NPP), estimating the effects of earthquakes has become more essential for the nuclear industry. Currently, most countermeasures against earthquakes are limited to large-scale disasters. Minor-scale earthquakes used to be ignored. Even though people do not feel the shaking due to minor earthquakes and minor earthquakes incur little damage to NPPs, they can change the environmental conditions, for instance, underground water level and the conductivity of the groundwater. This study conducted a questionnaire survey of residents living in the vicinity of an NPP to determine their perception and acceptance of plant safety against minor earthquakes. The results show that the residents feel earthquakes at levels that can be felt by people, but incur little damage to NPPs, as minor earthquakes (magnitude of 2.0–3.9) and set this level as a standard for countermeasures. Even if a minor earthquake has little impact on the safety of an NPP, there is still a possibility that public opinion will get worse. This study provides analysis results about problems of earthquake measures of Korean NPPs and specific things that can bring about an effect of deterioration of public acceptance. Based on these data, this article suggests that active management of minor earthquakes is necessary for the sustainability of nuclear energy.

Regulations on radioactive effluents released from Korean nuclear power plants

Nuclear power plants (NPPs) control the release of radioactive materials to the environment to protect both the members of the public and the biosphere from radiation exposure. Radioactive materials generated during the NPP operation are classified as gaseous, liquid, and solid wastes. Since solid radioactive wastes are normally kept in waste storage in NPPs and sent to a repository for disposal, their release to the environment is very unlikely. On the other hand, gaseous and liquid radioactive materials are released to the environment through discharge routes after adequate reduction and dilution processes. These radioactive materials are not only rigorously monitored by NPPs but also controlled by various regulations.

Effect of the Duration of Meteorological Data Collection on the Prospective Assessment of Long-Term Atmospheric Dispersion Factors.

Assessing the environmental dispersion of radioactive effluents from a nuclear facility under normal operation is a prerequisite in applying for the construction permit and operation license of the facility. In Korea, the Nuclear Safety and Security Commission requires at least 1- and 2-year periods of site-specific meteorological data to assess the atmospheric dispersion of the effluents from a nuclear reactor facility for a construction permit and operating license, respectively. In this study, we investigated if an extended period of meteorological data would be certainly preferable for the prospective assessment of dispersion of the atmospheric effluents under current data handling approaches. For test estimation, the meteorological data collected in the Kori nuclear power plant site in 2001 to 2008 were utilized. We suggest that the accumulated meteorological data be reflected in assessing the atmospheric dispersion in the form of individual annual data rather than the averaged data over the whole period.

Radioactive effluents released from Korean nuclear power plants and the resulting radiation doses to members of the public

Abstract Korean nuclear power plants (NPPs) periodically evaluate the radioactive gaseous and liquid effluents released from power reactors to protect the public from radiation exposure. This paper provides a comprehensive overview of the release of radioactive effluents from Korean NPPs and the effects on the annual radiation doses to the public. The amounts of radioactive effluents released to the environment and the resulting radiation doses to members of the public living around NPPs were analyzed for the years 2011–2015 using the Korea Hydro & Nuclear Power Co., Ltd's annual summary reports of the assessment of radiological impact on the environment. The results show that tritium was the primary contributor to the activity in both gaseous and liquid effluents. The averages of effective doses to the public were approximately on the order of 10 −3 mSv or 10 −2 mSv. Therefore, even though Korean NPPs discharged some radioactive materials into the environment, all effluents were within the regulatory safety limits and the resulting doses were much less than the dose limits.

Development of experimental metal sphere-mass map of reactor vessel scale models

Loose parts in coolant systems of a nuclear power plant (NPP) can damage structural integrity of the systems and loose part signals monitoring systems (LPMS) are installed in most of NPPs. LPMS detect impact signal of a loose part and generate alarm if amplitude of the signal is over preset level. If impact signal detected, mass estimation of the loose part must be carried out in order to see the effect of the part impact on structural integrity. A metal sphere-mass map, which shows relations between amplitude and center frequency of impact signals according to mass, is essential in mass evaluation of a loose part and the map built for US NPPs has been used so far. For the purpose of precise impact analysis suitable for Korean NPPs, this work discusses the development of a mass map of Korean Standard Nuclear Power Plant (KSNP) based on experimental data for scale models of a KSNP reactor vessel as tests for real NPP vessels under operation is actually impossible. Impact tests were carried out for scale models by using metal sphere with various mass and velocity. Amplitudes and center frequencies of impact signals were investigated based on time-frequency analysis and the mass map for each models were constructed based on the results. Blind tests were additionally implemented with metal sphere with unknown mass and velocity and verified feasibility of the proposed metal sphere mass map.